This invention relates to members made of a new austenitic steel which are irradiated with radio-active rays and are in contact with high-temperature, high-pressure water and, more particularly, to an austenitic steel which is advantageous especially in intergranular corrosion resistance and new nuclear reactors and nuclear fusion reactors made of this steel.
At present, structural members irradiated with neutrons in the core of a light-water nuclear reactor are made of the stainless steels SUS 304 and 316 specified in JIS. The chromium contained in stainless steels contributes to the improvement in intergranular corrosion resistance. However, neutron irradiation at high temperatures forms layers deficient in chromium near the grain boundaries; this results in the formation of what is called a sensitized structure, which increases sensitivity to intergranular corrosion or boundary-type stress corrosion cracking. Methods for reducing or preventing intergranular corrosion and boundary-type stress corrosion cracking (hereinafter collectively referred to as intergranular corrosion) in steel by improving the properties of austenitic stainless steels, are disclosed in Japanese Patent Examined Publication No. 1-18143, Japanese Patent Unexamined Patent Publication Nos. 62-238355 and 62-238353. These methods involve reducing the carbon contained in stainless steels and adding carbide stabilizing elements, such as niobium and titanium, to prevent the precipitation of chromium carbides.
The above-mentioned conventional arts were developed to suppress the formation of layers deficient in chromium near grain boundaries (what is called the sensitized structure), which is caused by the precipitation of chromium carbides, and to improve intergranular corrosion resistance thereby. In particular, these techniques are intended for stainless steels used in structural members of a core of a nuclear reactor and the first wall of a nuclear fusion reactor.
However, it has become evident that when stainless steels which have undergone solution treatment are irradiated with neutrons, layers deficient in chromium are formed near the grain boundaries along with the knocking-out of atoms from the base metal. Layers deficient in chromium are formed by neutron irradiation according to the following mechanism:
(1) Atoms of the base metal are knocked out into steel by neutron irradiation, and pairs of an interstitial atom and a vacancy are formed.
(2) Both the interstitial atom and the vacancy move through the steel. During this movement, they join a solute atom, causing selective diffusion of the solute atom.
(3) In the stainless steels of JIS SUS 304 and 316, neutron irradiation causes chromium to move away from grain boundaries, generating layers deficient in chromium at grain boundaries.
Further, the improvement of corrosion resistance of the JIS SUS 304 stainless steel for structural members of the core of a nuclear reactor is disclosed in Japanese Patent Unexamined Publication No. 62-107047 as follows:
(1) Neutron irradiation causes silicon and phosphorus to segregate at grain boundaries, reducing corrosion resistance.
(2) Neutron irradiation forms chromium carbides near grain boundaries, sensitizing the structure.
To prevent these phenomena, the following are disclosed in the above patent publication:
1) The silicon and phosphorus contents are lowered.
2) The formation of chromium carbides is suppressed by adding elements that form stable carbides, such as molybdenum, niobium and titanium.
It is, however, impossible to prevent the above-mentioned decrease in intergranular corrosion resistance caused by sensitization due to layers deficient in chromium formed by neutron irradiation, only by controlling concentrations of impurities.
In the above conventional techniques, consideration is not given to the phenomenon that layers deficient in chromium are formed at grain boundaries due to the knocking-out and diffusion of chromium by neutron irradiation, and corrosion resistance in water at high temperatures poses a problem.